Reprocessing Of Fuels Research Articles

Influence of the gamma source on the radiolytic stability of N,N,N′,N′-tetra-n-octyl-diglycolamide (TODGA)

Abstract In nuclear reprocessing, hydrolytic and radiolytic stability of ligands, extractants, diluents and solvents is of particular importance. The strength of extraction systems can only be fully exploited when process conditions are predictable and essential molecules, which are in close proximity to radioactive isotopes, are resistant to the radiation. During the development of novel extraction systems, testing of the radiolytic stability is often evaluated by exposing the molecules to high energetic gamma rays. In this work, the influence of the source of gamma rays is evaluated. TODGA was irradiated from 0–500 kGy using different sources of gamma rays, and its degradation was followed using UPLC-HRMS. Pool-type gamma irradiation facilities BRIGITTE B (Co-60) and GEUSE II (spent nuclear fuel) at SCK CEN were employed to serve as gamma sources. In this work, it is found that there is no significant influence on TODGA degradation which can be attributed to the origin of the gamma source.

Corrosion-mediated production of uranium(III) chloride from metallic uranium in molten LiCl–KCl salt contained within a stainless-steel crucible

Uranium (III) chloride (UCl3) is a crucial component of a potent nuclear recycling technology—pyroprocessing—and next-generation molten salt reactors. It is usually synthesized by reacting metallic uranium with chlorinating agents (e.g., CdCl2 and PbCl2) in molten chloride salts. In this study, we report the unexpected formation of UCl3 from metallic simulated fuel (simfuel) immersed in impure molten LiCl–KCl salt (in the presence of a small amount of residual H2O) in a stainless-steel (SS) crucible, without a chlorinating agent. We investigated various factors influencing UCl3 formation, including fuel type (metallic simfuel, pure U, oxide simfuel, or no fuel), crucible material (SS or alumina), salt composition (LiCl–KCl or LiCl), temperature (773 K or 923 K), and contact between fuel and SS crucible. UCl3 only formed when metallic fuels (simfuel or pure U) were immersed in molten salt in the SS crucible, with higher concentrations at elevated temperatures. Oxide fuels did not produce UCl3, nor did contact with the crucible affect formation. Our findings suggest that impurities, particularly moisture in the salt, corroded the SS crucible, releasing iron and chromium chlorides that reacted with metallic U to form UCl3. UCl3 formation was more pronounced in LiCl–KCl than in LiCl, and thermodynamic calculations helped establish the mechanism.

Anthropocene 129I Record in the Yellow Sea Sediments and Its Indication for River-Delivered Radioactive Pollution to Marginal Seas.

As the number of coastal nuclear facilities rapidly increases and the wastewater from the Fukushima Nuclear Plant has been discharged into the Pacific Ocean, the nuclear environmental safety of China's marginal seas is gaining increased attention along with the heightened potential risk of nuclear accidents. However, insufficient work limits our understanding of the impact of human nuclear activities on the Yellow Sea (YS) and the assessment of their environmental process. This study first reports the 129I and 127I records of posthuman nuclear activities in the two YS sediments. Source identification of anthropogenic 129I reveals that, in addition to the gaseous 129I release and re-emission of oceanic 129I discharged from the European Nuclear Fuel Reprocessing Plants (NFRPs), the Chinese nuclear weapons testing fallout along with the global fallout is an additional 129I input for the continental shelf of the YS. The 129I/127I atomic ratios in the North YS (NYS) sediment are significantly higher than those in the other adjacent coastal areas, attributed to the significant riverine input of particulate 129I by the Yellow River. Furthermore, we found a remarkable 129I latitudinal disparity in the sediments than those in the seawaters in the various China seas, revealing that sediments in China's marginal seas already received a huge anthropogenic 129I from terrigenous sources via rivers and thus became a significant sink of anthropogenic 129I. This study broadens an insight into the potential impacts of terrigenous anthropogenic pollution on the Chinese coastal marine radioactive ecosystem.

Densities and Molar Volumes for Binary Solutions of Tri-iso-Amyl Phosphate and n-Dodecane in the Temperature Range of 283.15–363.15 K

Tri-n-butyl phosphate (TBP) has remained workhorse of PUREX based nuclear recycle operations since 1956. However due to its associated disadvantages like third phase formation and tendency to react violently with nitric acid at elevated temperatures generated requirement of an alternate extractant for PUREX process. Tri-iso-amyl phosphate (TiAP) has better attributes than TBP on both the counts. Density of solutions is an important process parameter for design of equipment, process operations and control and safety analysis. In this work, experimental data on the density of binary solution of tri-iso amyl phosphate and n-dodecane are reported for several temperature levels. The related volumetric parameters were also estimated from experimental data as well as Kirkwood–Buff integrals. Derivative parameters like linear coefficients of preferential solvation δij(i≠j)o\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$\\delta_{ij(i \ e j)}^{o}$$\\end{document} and local mole fractions xij(i≠j)\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$x_{ij(i \ e j)}$$\\end{document} at 298.15 K were also estimated and reported.Graphical

Synthesis, Characterization and Dissolution behavior of Nitrides of U-Zr and U-Pu-Zr metallic alloy fuels for Aqueous reprocessing

The aqueous based PUREX process is an alternate choice to pyro-reprocessing of metallic alloy fuels (U-Zr/U-Pu-Zr). Direct dissolution of metallic alloy fuels in a nitric acid medium is not feasible due to vigorous reaction exhibited by the alloys when exposed to this medium. Conversion of metallic alloys into their oxides is a challenging method due to poor dissolution of oxides in nitric acid. Excellent dissolution of nitrides compared to oxides of metallic alloy fuels lead to exploit the nitride route for aqueous reprocessing. Hence, a novel method was adopted to prepare the nitrides of unirradiated U-Zr and U-Pu-Zr alloys via hydridation using UH3, as source-cum-getter instead of hydrogen cylinder. In this context, nitrides of unirradiated U-Zr/U-Pu-Zr metallic alloys were prepared by nitridation via hydridation. Initially, nitrides of U and Zr metals were synthesized and the formation of ZrN and U2N3 and UN phases were confirmed by powder X-ray diffraction measurements. Subsequently, nitrides of unirradiated U-Zr and U-Pu-Zr alloys were synthesized and characterized by X-ray diffraction, scanning electron microscopy and particle analyzer. The dissolution behavior of nitrides of Zr, U, U-Zr and U-Pu-Zr alloys were examined in nitric acid (12 M) medium under reflux conditions at 383 K for 8 h. The present study highlights the potential of the nitride route for efficient dissolution of actinides (U and Pu) and it is very important to develop flow sheets for the aqueous reprocessing of metallic alloys fuels.

A self-optimised approach to synthesising DEHiBA for advanced nuclear reprocessing, exploiting the power of machine-learning

To aid the advancement of hydrometallurgical reprocessing of used nuclear fuel, this work has explored and optimised the synthesis of DEHiBA in continuous flow, to establish a scalable, cost-effective manufacture route.

Evaluation of fracture resistance behavior of Alloy 690 material upto 1000°C

Alloy 690 is a nickel-based alloy with a nominal composition of 30% Cr and 10% Fe (wt%). It is denoted as Ni-30Cr-10Fe. This alloy is currently being investigated by various researchers for its use in nuclear industry as a possible replacement of alloy 600 steam generator tubes because of its excellent resistance to inter-granular stress corrosion cracking in water reactor environment. Alloy 690 is mainly used in vitrification melters for immobilization of high level waste in nuclear recycle plant. The temperature of water reactor environment is considerably lower compared to that of vitrification application, where temperature can reach an order of 1000 to 1100˚C. The mechanical and fracture properties of alloy 690 at the elevated temperature is important from the point of view of design and safety analysis of vitrification melter components and it is not available in literature. For this purpose, single edged notched tensile specimens have been fabricated from alloy 690 plates of 10 mm thickness. The fracture specimens have been tested in the temperature range of 25˚C to 1000˚C, which is the operating range of vitrification melters. The crack initiation toughness was obtained using modified blunting line equation in order to account for material strain hardening effect at a given temperature. From the fracture test results, it was observed that the initiation toughness is lowest at 700˚C. To understand this behavior, scanning electron microscopy was carried out on the fracture surface of the tested specimens. The fracture surface of the specimen tested at 700˚C showed presence of brittle phases in the microstructure, which are responsible for lower value of fracture toughness at this condition. The results of this work shall be useful for design and analysis of vitrification melter components.

Selective Separation of Zr(IV) from Simulated High-Level Liquid Waste by Mesoporous Silica.

The efficient separation of Zr(IV) ions from strong acidic and radioactive solutions is a significant challenge, especially in the context of the aqueous reprocessing of nuclear fuels. The complexity of such solutions, which are often characterized by high acidity and the presence of radioactive elements, poses formidable challenges for separation processes. Herein, several mesoporous silicas (HMS, MCM-41, KIT-6, and SiO2-70 Å) with excellent acid and radiation resistance properties were employed as sorbents to remove Zr(IV) ions from simulated high-level liquid waste. The batch experiments were designed to investigate the influence of adsorption time, HNO3 concentration, initial Zr(IV) concentration, adsorbent dosage, and temperature on the adsorption behavior of Zr(IV). The results indicate that the adsorption equilibrium time of mesoporous silica materials was approximately 8 h, and all the adsorption processes followed the pseudo-second-order kinetics equation. The isotherms of Zr(IV) adsorption by KIT-6 exhibited good agreement with the Langmuir model, while the Freundlich model could be utilized to fit the adsorption on HMS, MCM-41, and SiO2-70 Å. The adsorption capacity of MCM-41 for Zr(IV) in 3 mol/L HNO3 was 54.91 mg/g, which is three times the adsorption capacity reported for commercial silica gel (17.91 mg/g). The thermodynamic parameters indicate that the adsorption processes for zirconium are endothermic reactions. Furthermore, the mesoporous silicas exhibited a pronounced selectivity in the adsorption of Zr(IV) within a simulated high-level liquid waste containing 10 co-existing cations (3 mol/L HNO3). This suggests that mesoporous silicas have great potential for Zr(IV) removal in actual radioactive liquids with high acidity during spent fuel reprocessing.

Reconstruction of the Sources and Their Contributions to 129I in Northern Xinjiang, China.

Anthropogenic 129I, as a long-lived fission product and volatile radionuclide, can be used to investigate dispersion of air masses and the deposition of atmospheric pollution. Surface soil and soil core samples were collected from Northern Xinjiang and analyzed for 127I and 129I. The results show that 129I/127I atomic ratios in surface soil are inhomogeneous with a range of (2.07-106) × 10-9, and the maximum values in each soil core occurred at surface-subsurface layers (0-15 cm) at undisturbed sites. The dominant source of 129I in Northern Xinjiang is European nuclear fuel reprocessing plant (NFRP) releases, accounting for at least 70% of the total inventory; less than 20% of 129I originates from the global fallout of atmospheric nuclear weapons tests; less than 10% comes from the regional deposition of nuclear weapons tests at the Semipalatinsk site; and the regional deposition from the nuclear weapons tests at the Lop Nor site is insignificant. The European NFRP-derived 129I was transported to Northern Xinjiang via long-distance atmospheric dispersion with the westerlies through Northern Eurasia. The distribution of 129I in the surface soil in Northern Xinjiang is mainly controlled by topography, wind fields, land utilization, and vegetation coverage.

Simultaneous determination of flouride, iodate and iodide by ion chromatography in caustic off-gas scrubber of nuclear recycle facilities by employing dual detectors

ABSTRACT An ion chromatography method was developed for the simultaneous determination of fluoride, iodate and iodide in a caustic scrubber solution used for trapping off-gases released during the dissolution of nuclear fuels. A separation between iodate and fluoride could be realized with 0.6 mM hydroxide, which is too weak to elute other anions including iodide in reasonable time. Studies were carried out at different column temperature, and with the help of Van’t Hoff plots the thermodynamic parameters associated with the column equilibrium were estimated, which helped in understanding the co-elution of iodate and fluoride at higher concentration of eluent. An arrangement of conductivity detector and UV-VIS detector in sequence after the ion exchange column (with 15 mM NaOH as eluent) was used for recording chromatograms for conductivity signal and UV-VIS signal simultaneously. The information from the two chromatograms was utilized to quantify iodate, fluoride and iodide in a single run. The limit of detection for fluoride, iodate and iodide is 7 ppb, 10 ppb and 12 ppb, respectively. The precision of the method is better than 2% (RSD) at 0.5 mg/L. The method was used successfully for the real sample analysis.

Vibration Analysis and Diagnostics of a High Speed Centrifuge for Fast Reactor Fuel Reprocessing Applications

This study aims at understanding the cause of vibrations observed during the operation of high speed centrifuges in a radioactive environment and materializing ways to mitigate these vibrations to ensure dynamic stability of centrifuges during operation. For this, a condition-based monitoring study was carried out for the centrifuge and its supporting structure using dynamic co-relation study in frequency domain during the operation of centrifuge at its operating speed of 19[Formula: see text]700[Formula: see text]rpm. Dynamic characteristics of the support structure were analyzed using standard finite element code. It was observed that high vibration amplitudes in the original support structure, before modification, owed to its low structural stiffness. Hence, structural modifications were carried out to reduce the dynamic crossover response of the support structure at the critical speeds and operating speeds of centrifuge. It was found that structural modification resulted in considerable reduction in the vibration amplitude of the support structure, compared to its initial configuration.

Identifying, Predicting and Preventing Localized Corrosion in Kr-85 Storage Canisters from the Reprocessing of Nuclear Fuel

During the reprocessing of nuclear fuel radioactive gases are released from the fuel matrix. These gases need to be separated and stored to prevent release into the environment. It is proposed that certain risk factors, which are elucidated herein, caused accelerated corrosion of the interior canister walls breaking containment of legacy storage canisters. The main risk factor of concern is the effect of Rb-85. Rb-85 is produced from the radioactive beta minus decay of Kr-85. The main question is whether an environment containing Rb as a decay product causes corrosion of the canister materials at an accelerated rate. If this does occur, the the corrosion threat will grow with time as more rubidium is produced.Based on the environment within the canister, and the risk factors of the system, a corrosion lifecycle has been developed and is briefly discussed. Stage 1 represents the canister at high temperature of 600°C after it has been sealed and hot isotactic pressed (HIP) for 4 hours. In this stage it is possible for oxides to form on the canister walls due to the residual oxygen in the canister from loading. It is possible that the presence of Rb will interact with the canister material and disrupt the formation of iron and chromium oxides. These Rb oxides, if present, may not inhibit corrosion thus causing an increase in the corrosion rates. Stage 2 represents the canister cooling from 600°C to ambient. In this stage wet thermal oxides will form on the canister walls as the temperature drops below the deliquescence point and water droplets appear, setting up the conditions for aqueous corrosion, like stage 1 the possible formation of RbxCryOzwhich may not be as protective as other oxides, can cause an increase in corrosion rates. Stage 3 represent oxide and environmental interactions. In this stage the environmental risk factors, such as the presence of cathodic reactants oxygen, hydron and peroxide, will create conditions to enable aqueous corrosion. Stages 4 through 8, describe the stages of pitting, uniform corrosion, and galvanic corrosion. Because the material chosen for the canister is 4130 low alloy steel, and the welding material used was 304 stainless steel, there is the possibility of galvanic corrosion. This can increase the uniform corrosion rate, and lead to more pitting events causing depassivation of the surface, or deeper pit penetration. The chosen canister materials are also at risk of pitting attack, the risk of which is increased as the canister is likely driven to alkalinity from the products of the cathodic reaction. Pitting is especially dangerous to the canister which needs to maintain containment for 100 years.This work looks to test the risk factors and develop a mechanistic understanding of the canister’s corrosion process with the goal of recommending a material that will be able to maintain containment for the needed 100 years if nuclear fuel were to start being repressed in the United States.

Gamma and beta radiolysis of tri-iso-amyl phosphate: Degradation of tri-iso-amyl phosphate and formation of di-iso-amyl phosphoric acid

Tri-iso-amyl phosphate (TiAP) is considered a potential alternative solvent to tri-butyl phosphate (TBP) for aqueous reprocessing of molten salt reactors and fast reactor fuels. This study explored the γ and β radiolysis of TiAP under various conditions through the analysis of the radiolytic concentration changes of TiAP and the yields of di-iso-amyl phosphoric acid (DiAP) using gas chromatography. The results indicated that the effects of dose, diluent, and nitric acid on the radiolysis of TiAP under γ-and β-irradiations showed a similar regularity. The degradation rate of TiAP followed a first-order reaction, and the paraffinic diluent was able to sensitize TiAP. Although the presence of nitric acid promoted the radiolysis of TiAP and the formation of DiAP in a neat TiAP system, it did not lead to significant changes in the 36%TiAP/n-DD system, implying that nitric acid is involved in an intricate process. Furthermore, based on these results, possible degradation mechanisms of TiAP were described.

Preparation of a chromium carbon coating on 316L by a disproportionation reaction in molten salts and its corrosion resistance in molten fluorides

Molten fluorides have been widely used in specific nuclear reprocessing and molten salt reactors due to their unique pyrophysical and pyrochemical properties. However, the severe corrosion of structural materials in highly corrosive molten fluorides at high temperatures limits their commercial use. In view of the high thermodynamic stability of chromium carbides in molten fluoride salts, and their good Cr-diffusion-blocking ability, it is proposed to develop a CrxCy coating, by electrodeposition coupled with a disproportionation reaction in molten salts, for application in molten fluorides A carbon coating is first prepared on 316L stainless steel by electrodeposition in molten LiCl-KCl-K2CO3, followed by a disproportionation reaction of Cr2+ on the carbon coating to obtain a CrxCy coating. Furthermore, the growth mechanism of the CrxCy coating and its corrosion resistance in molten fluorides are also investigated.

Fate of Neptunium in nuclear fuel cycle streams: state-of-the art on separation strategies

Abstract Neptunium, with a half life of 2.14 million years is one of the most notorious activation products in the nuclear fuel cycle. It has been more than 5 decades in the reprocessing of nuclear fuels by the well documented PUREX process, but the fate of Np in the PUREX cycle is still not well controlled. Although Np being stable in its pentavalent state in low acid media, its starts to undergo disproportionation at higher acidities. This disproportionation along with the oxidizing conditions of the HNO3 medium makes Np to exits as Np(IV), Np(V) and Np(VI) in the dissolver solution. The overall extractability of Np in the co-decontamination step of the PUREX cycle is dependent on its oxidation state in the medium as Np(VI) and Np(IV) being extractable while Np(V) being least extractable. The present review article discusses about the speciation of Np in HNO3 and its disproportionation. The variety of redox reagents are discussed for their effectiveness towards controlling Np redox behavior in the HNO3 media. The extraction of Np with the different class of extractant has also been discussed and the results are compared for better understanding. Solid phase extraction of Np using both commercially available resin and lab based synthesized resins were discussed. The anion exchange resins with the different cationic centers were shown to behave differently towards the uptake of Np form the acidic medium. The present review also highlight the chemical conditions required for controlling or minimizing the fate of Np in different process streams of the nuclear fuel cycle.

Development of monazite glass-ceramic wasteforms for the immobilisation of pyroprocessing wastes

Pyrochemical reprocessing is a potential route for the reprocessing of fuels from next-generation reactors. Lanthanide fission products may be separated from the reprocessing salt by a number of methods; however, salts may be entrained in the resultant product. This work demonstrates conceptual monazite glass-ceramic wasteforms based on the quarternary (100−x)(10Na2O–36Fe2O3–54P2O5)–xLa2O3 (x = 5, 10, 15), with La2O3 as a surrogate for the lanthanide oxide waste stream. Samples were produced by the melting of La2O3 with a sodium-iron-phosphate glass frit and characterised by XRD and SEM–EDX. The monazite phase was successfully formed at all waste loadings with clear segregation of La to the crystalline phase; however, high La2O3 loading (>10 mol%) was found to destabilise the glass system resulting in gross crystallisation. These initial results indicate that monazite glass-ceramics are promising wasteforms for this waste stream.

Nonlinear Multivariate Regression Algorithms for Improving Precision of Multisensor Potentiometry in Analysis of Spent Nuclear Fuel Reprocessing Solutions

Potentiometric multisensor systems were shown to be very promising tools for the quantification of numerous analytes in complex radioactive samples deriving from spent nuclear fuel reprocessing. Traditional multivariate calibration for these multisensor systems is performed with partial least squares regression—an intrinsically linear regression method that can provide suboptimal results when handling potentiometric signals from very complex multi-component samples. In this work, a thorough investigation was performed on the performance of a multisensor system in combination with non-linear multivariate regression models for the quantification of analytes in the PUREX (Plutonium–URanium EXtraction) process. The multisensor system was composed of 17 cross-sensitive potentiometric sensors with plasticized polymeric membranes containing different lipophilic ligands capable of heavy metals, lanthanides, and actinides binding. Regression algorithms such as support vector machines (SVM), random forest (RF), and kernel-regularized least squares (KRLS) were tested and compared to the traditional partial least squares (PLS) method in the simultaneous quantification of the following elements in aqueous phase samples of the PUREX process: U, La, Ce, Sm, Zr, Mo, Zn, Ru, Fe, Ca, Am, and Cm. It was shown that non-linear methods outperformed PLS for most of the analytes.

A Homocoupling Approach to the Key Dione of CyMe4-BTPhen – Vital Ligands for Nuclear Clean-Up by the SANEX Process

AbstractCyMe4-BTPhen and CyMe4-BTBP are the principal ligand systems used in Europe for the separation of actinides from lanthanides as a part of the SANEX process for nuclear recycling and reprocessing. We present a new approach to the synthesis of the CyMe4 fragment beginning from readily available hydroxypivalic acid. It features a cobalt-catalysed homocoupling of a neopentyl bromide to provide the key bisester precursor, thereby avoiding the requirement for technically challenging low-temperature LDA-mediated aldol chemistries.

Physical Aspects of the Waste of Re-Enrichment Utilization for Recycled Uranium from VVER-1000/1200 Spent Fuel Reprocessing in Molten Salt Reactors

Physical Aspects of the Waste of Re-Enrichment Utilization for Recycled Uranium from VVER-1000/1200 Spent Fuel Reprocessing in Molten Salt Reactors

Deciphering the curved profile of uranyl ions at the aqueous-organic interface by atomistic simulations

The acceptance of nuclear energy is due to the safe and clean operation of the reactor along with the proper reprocessing of the aqueous nuclear waste. In that context, extraction of uranyl ion form aqueous solution has an immense importance in the field of fuel reprocessing. The extraction of uranyl ion is the key step on which the efficiency of the reprocessing depends. Tri-isoamyl phosphate (TiAP) is proven to be a potential alternative of Tri-butyl phosphate (TBP) in the Plutonium Uranium Extraction (PUREX) process. Therefore, understanding of PUREX process at the atomistic level using molecular dynamics simulations is of immense practical significance. Therefore, extensive molecular dynamics simulations have been conducted using calibrated all-atom optimized potentials. The simulation studies have been performed for a wide range of nitric acid concentration from 1 M to 8 M. MD studies reveal the importance of interfacial properties on the extraction of uranyl ions. The predicted interface thickness is seen to be increased with acidity whereas the calculated interfacial tension shows the opposite behavior. From the calculated results, it is observed that the capillary part of the total interface thickness is suppressed whereas the intrinsic part is expanded due to the packing of uranyl ions. The uranyl complexes remain adsorbed at the close proximity of the aqueous-organic interface. The predicted distribution constant of uranyl ion with increasing acidity exhibits a “curved profile“ as observed in the experiment. The computed interaction energy using DFT displays that the TiAP-UO22+ interaction energy is much higher compared to the TiAP-HNO3 and TiAP-H2O interaction energies thus support the higher distribution constant of uranyl ion. The high value of negative free energy difference, ΔΔG (ΔGSolvation - ΔGHydration) of uranyl ion in the organic and aqueous phase confirms that the partitioning of uranyl ion to the organic phase through complexation reaction is thermodynamically feasible process.